Numerical Evaluation of Fluid Mixing Phenomena in Boiling Water Reactor Using Advanced Interface Tracking Method

Yoshida, Hiroyuki; Takase, Kazuyuki

doi:10.1299/jfst.3.311

Cited by 3 publications

(1 citation statement)

References 6 publications

(8 reference statements)

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“…In this study, to develop the prediction technology of twophase flow dynamics under earthquake conditions, a detailed two-phase flow simulation code with an advanced interface tracking method TPFIT [2] was expanded to two-phase flow simulation under earthquake conditions. In this technology development, database based on experimental results is indispensable to validate expanded simulation code [3,4].…”

Section: Introductionmentioning

confidence: 99%

Numerical Simulation of Two-Phase Flow Behavior under Earthquake Acceleration

Yoshida

Takase

Kaneko

et al. 2012

2012 Asia-Pacific Power and Energy Engineering Conference

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Fluctuation of void faction is an important factor for the safety operation of the nuclear reactor, and fluctuation of void fraction is related to behavior of gas-liquid two-phase flow. In case of the earthquake, the fluctuation is not only the flow rate, but also body force on the two-phase flow and shear force through a pipe wall. Interactions of gas and liquid through their interface also act on the behavior of the two-phase flow. The fluctuation of the void fraction is not clear for such complicated situation under the earthquake. Therefore, the behavior of gasliquid two-phase flow is investigated experimentally and numerically in a series of study. In this study, to develop the prediction technology of two-phase flow dynamics under earthquake acceleration, a detailed two-phase flow simulation code with an advanced interface tracking method TPFIT was expanded to two-phase flow simulation under earthquake conditions. The bubbly flow in a horizontal pipe excited by oscillation acceleration and under the fluctuation of the liquid flow was simulated by using the expanded TPFIT. And the influence of acceleration, the fluctuation of flow rate and the fluctuation of share force though a pipe wall were investigated in the numerical simulations.

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Section: Introductionmentioning

confidence: 99%

Numerical Simulation of Two-Phase Flow Behavior under Earthquake Acceleration

Yoshida

Takase

Kaneko

et al. 2012

2012 Asia-Pacific Power and Energy Engineering Conference

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Thermal-hydraulic analysis of light water reactors under different steady-state operating conditions, Part 1: Boiling water reactor

Hutli

Kridan

2022

NUCL TECH RAD PROT

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The steady-state thermal-hydraulic analysis of the core of the Boiling Water Reactor (BWR/6) at nominal operating conditions is presented in this paper. The BWR/6 is produced by General Electric USA. The analysis' goal is to keep the thermal safety margin under control and the core integrity intact under steady-state operating conditions. The effects of operating conditions such as power distribution, power level, and coolant mass flow rate on the pro- posed core's performance are investigated. For this purpose, the one-dimensional computer code MITH was used. The code's reliability was tested using the General Electric benchmark 3579 MW reactor. Two-channel models were tested (the average and the hot channel). Ther- mal-hydraulic parameters such as fuel-centerline, fuel-surface, outer clad surface and coolant temperature, critical and actual local heat flux, critical and minimum critical heat flux ratio and pressure drop are evaluated along the tested channels. Temperatures, as well as actual and critical heat flux distribution profiles, were obtained. The tested operating conditions had a significant influence on these parameters, and also on the thermal-hydraulic performance. The obtained results are in good agreement with the data from the tested core. The obtained results are well within the safety margins. The good agreement between tested reactor data and MITH code calculation concerning the reactor demonstrates the reliability of the analysis methodology from a thermal-hydraulic perspective.

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Thermal-hydraulic analysis of light water reactors under different steady-state operating conditions, Part 2: Pressurized water reactor

Hutli

Kridan

2022

NUCL TECH RAD PROT

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The 1-D computer code MITH was used in this paper to perform sub-channel thermal-hydraulic analyses of a typical (Westinghouse model) pressurized water reactor. Two typical channels, hot and average, with the same flow rate and pressure drop, were tested under steady-state operating conditions. In this analysis, the channel with the highest temperature is designated as the hot channel. For the calculations, the channel model was divided into 20 parts. The thermal-hydraulic performance of the tested reactor was affected by power distribution, power level, and coolant mass-flow rate. Temperature distribution profiles of the fuel element and coolant are obtained for the average and hottest channels. A critical heat flux qncr analysis is also carried out and the heat fluxes in both channels were calculated. The W-3 correlation is employed to examine qncr in the hottest channel. Some data from the pressurized water reactor typical data sheet were used as input data, while others were used to validate the code. The code faithfully reproduced the Westinghouse model reactor results, including coolant, cladding, centerline, and surface fuel temperatures, quality and local heat flux qnloc, qncr and minimum departure from nucleate boiling ratio.

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Numerical Evaluation of Fluid Mixing Phenomena in Boiling Water Reactor Using Advanced Interface Tracking Method

Cited by 3 publications

References 6 publications

Numerical Simulation of Two-Phase Flow Behavior under Earthquake Acceleration

Numerical Simulation of Two-Phase Flow Behavior under Earthquake Acceleration

Thermal-hydraulic analysis of light water reactors under different steady-state operating conditions, Part 1: Boiling water reactor

Thermal-hydraulic analysis of light water reactors under different steady-state operating conditions, Part 2: Pressurized water reactor

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